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JAEA Reports

Safety analysis for severe accidents anticipated during truck transport of uranium fresh fuel and raw material (Contract research)

Nomura, Yasushi*; Takahashi, Satoshi*; Okuno, Hiroshi

JAEA-Technology 2008-009, 273 Pages, 2008/03

JAEA-Technology-2008-009.pdf:8.92MB

Safety demonstration analyses were conducted under contract with the Ministry of Economic, Trade and Industry of Japan from 2001 to 2004 for the purpose of assuaging public jitters concerning the transport. The current transport routes and the past accident/incident records were surveyed, three accident scenarios, i.e., a fall from an overpass, an open fire after collision with an oil tank trailer, a fire caused by collision with 2-ton truck inside a tunnel were set up. Mechanical damages and thermal failures were analyzed using the finite element codes LS-DYNA and ABAQUS. In addition, criticality safety analyses were made using the continuous energy Monte Carlo code MVP for the transport casks damaged in reference to the previous mechanical and thermal analyses. Thus, the integrity of packaging against leakage of radioactive material was shown in the case of severe accidents anticipated to occur during transportation without any harmful effect to the public and environment.

Journal Articles

A Modular metal-fuel fast reactor with one-loop main cooling system

Chikazawa, Yoshitaka; Okano, Yasushi; Konomura, Mamoru; Sato, Koji; Sawa, Naoki*; Sumita, Hiroyuki*; Nakanishi, Shigeyuki*; Ando, Masato*

Nuclear Technology, 159(3), p.267 - 278, 2007/09

 Times Cited Count:1 Percentile:86.85(Nuclear Science & Technology)

A diversified or modular power source is attractive since it requires a low construction cost per unit and can be demonstrated in small scale experimental facilities. In this study, a new metal fuel sodium cooled reactor with 300MW electric has been developed enhancing cost reduction. And economical potential at demonstration stage with first of a kind (FOAK) is emphasized. A minimum configuration with a compact reactor vessel, a one-loop main cooling system and a simple fuel handling system is adopted enhancing cost reduction. For safety evaluation, reliability of the one-loop main cooling system has been shown by pipe-break transient analyses. Besides, construction cost of a demonstration plant with a first reactor and a small reprocessing and fuel fabrication facility is also evaluated. A major feature of the present concept is that the demonstration reactor and facilities can be directly appropriated for first commercial modules and the power plant can easily increase its capacity adding reactor and electrorefiner modules. A fast reactor cycle commercialization scenario using the present concept is thought to give low R&D or investment risk and high cost performance since the total demonstration plant cost is relatively small and the facilities are directly appropriated to commercial use.

Journal Articles

A Compact loop-type fast reactor without refueling for a remote area power source

Chikazawa, Yoshitaka; Okano, Yasushi; Konomura, Mamoru; Sawa, Naoki*; Shimakawa, Yoshio*; Tanaka, Toshihiko*

Nuclear Technology, 157(2), p.120 - 131, 2007/02

 Times Cited Count:1 Percentile:86.85(Nuclear Science & Technology)

A small reactor has a potential to be utilized as a power source applicable to diversified social needs and reduce capital risks. In remote sites where the population is small and plants can not be economically connected to a power grid, power sources without refueling whose capacities are lower than 50 MWe are required because fuel transfer cost is expensive in such sites. In the present study, a small sodium cooled core with 30 years lifetime has been developed and a simple plant system without refueling has been sketched. Dimensions of major components are determined to evaluate its economical potential. Transient analyses show that self actuated shutdown system (SASS) enhances the passive safety features to maintain the reactor integrity in anticipate transient without scram events.

Journal Articles

A Modular metal fuel fast reactor enhancing economic potential

Chikazawa, Yoshitaka; Okano, Yasushi; Konomura, Mamoru; Sato, Koji; Ando, Masato*; Nakanishi, Shigeyuki*; Sawa, Naoki*; Shimakawa, Yoshio*

Proceedings of 2006 International Congress on Advances in Nuclear Power Plants (ICAPP '06) (CD-ROM), 8 Pages, 2006/06

A diversified or modular power source is attractive since it requires a low construction cost per unit and can be demonstrated in small scale experimental facilities. In this study, a new metal fuel sodium cooled reactor with 300MW electric has been developed enhancing cost reduction. And economical potential at demonstration stage with first of a kind (FOAK) is emphasized. A minimum configuration with a compact reactor vessel, a one-loop main cooling system and a simple fuel handling system are adopted enhancing cost reduction within safety requirement. Besides, construction cost of a demonstration plant with a first kind of reactor and a small reprocessing and fuel manufacturing facility is also evaluated. A major feature of the present concept is that the demonstration facilities can be appropriated for commercialized ones since they can be easily commercialize by increasing reactor and electrorefiner modules. A FBR cycle commercialization scenario using the present concept is thought to give low risk and high cost performance since the total demonstration plant cost is relatively small and the facilities are directly appropriated to commercial use.

Journal Articles

Conceptual design study of helium cooled fast reactor in the "feasibility study" in Japan

Okano, Yasushi; Naganuma, Masayuki; Ikeda, Hirotsugu; Mizuno, Tomoyasu; Konomura, Mamoru

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10

Conceptual design of gas-cooled fast reactor (GFR) have been studied for selecting applicable combinations of coolant, fuel material and configuration, and, balance of plant, as a part of feasibility study in Japan. Large-scale He-cooled GFR, employing mixed nitride fuel and achieving a high core outlet temperature of 850$$^{circ}$$C, is recognized to achieve attractive features as a future nuclear reactor system. Three fuel configurations are considered and compared in their core and safety performances; one is horizontal-flow cooling fuel assembly (F/A), another is hexagonal matrix block F/A, and the last one is sealed pin bundle F/A. The horizontal-flow and matrix block F/A cores show nearly the same neutronics performances on discharge burnup around 120GWd/ton, breeding ratio above 1.1, and, core cooling performances under depressurization condition without control rod scram or auxiliary core cooling system (ACCS) actuations; whereas around 30% smaller quantity of fissile Pu required is a merit for matrix concept. The sealed pin bundle F/A core potentially shows attractive neutronics performances on discharge burnup about 141GWd/ton with breeding ratio of 1.27, although rapid control rod scram and ACCS actuations are indispensable for core cooling under depressurization accident conditions.

Journal Articles

Validation of simplified evaluation models for first peak power, energy, and total fissions of a criticality accident in a nuclear fuel processing facility by TRACY experiments

Nomura, Yasushi*; Okuno, Hiroshi; Miyoshi, Yoshinori

Nuclear Technology, 148(3), p.235 - 243, 2004/12

 Times Cited Count:3 Percentile:73(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Feasibility Study on Commercialization of Fast Breeder Reactor Cycle Systems Interim Report of Phase II; Technical Study Report for Reactor Plant Systems

Konomura, Mamoru; Ogawa, Takashi; Okano, Yasushi; Yamaguchi, Hiroyuki; Murakami, Tsutomu; Takaki, Naoyuki; Nishiguchi, Youhei; Sugino, Kazuteru; Naganuma, Masayuki; Hishida, Masahiko; et al.

JNC-TN9400 2004-035, 2071 Pages, 2004/06

JNC-TN9400-2004-035.pdf:76.42MB

The attractive concepts for Sodium-, lead-bismuth-, helium- and water-cooled FBRs have been created through using typical plant features and employing advanced technologies. Efforts on evaluating technological prospects of feasibility have been paid for these concepts. Also, it was comfirmed if these concepts satisfy design requierments of capability and performance presumed in the feasibilty study on commertialization of Fast Breeder Reactor Systems. As results, it was concluded that the selection of sodium-cooled reactor was most rational for practical use of FBR technologies in 2015.

JAEA Reports

Criticality safety assessment by assuming spent fuel burnup distribution; Examination of various methods for setting burnup, 1 (Contract research)

Nomura, Yasushi*; Okuno, Hiroshi; Miyoshi, Yoshinori

JAERI-Tech 2004-030, 64 Pages, 2004/03

JAERI-Tech-2004-030.pdf:4.59MB

no abstracts in English

Journal Articles

Revised data for 2nd version of Nuclear Criticality Safety Handbook/Data Collection

Okuno, Hiroshi; Ryufuku, Susumu*; Suyama, Kenya; Nomura, Yasushi; Tonoike, Kotaro; Miyoshi, Yoshinori

JAERI-Conf 2003-019, p.116 - 121, 2003/10

This paper outlines the data prepared for the 2nd version of Data Collection of the Nuclear Criticality Safety Handbook. These data are discussed in the order of its preliminary table of contents. The nuclear characteristic parameters (k$$_{rm inf}$$, M$$^{2}$$, D) were derived, and subcriticality judgment graphs were drawn for eleven kinds of fuels which were often encountered in criticality safety evaluation of fuel cycle facilities. For calculation of criticality data, benchmark calculations using the combination of the continuous energy Monte Carlo criticality code MVP and the Japanese Evaluated Nuclear Data Library JENDL-3.2 were made. The calculation errors were evaluated for this combination. The implementation of the experimental results obtained by using NUCEF facilities into the 2nd version of the Data Collection is under discussion. Therefore, related data were just mentioned. A database is being prepared to retrieve revised data easily.

Journal Articles

Improvements to SFCOMPO; A Database on isotopic composition of spent nuclear fuel

Suyama, Kenya; Nouri, A.*; Mochizuki, Hiroki*; Nomura, Yasushi*

JAERI-Conf 2003-019, p.890 - 892, 2003/10

Isotopic composition is one of the most relevant data to be used in the calculation of burnup of irradiated nuclear fuel. Since autumn 2002, the Organisation for Economic Co-operation and Development/Nuclear Energy Agency OECD/NEA) has operated a database of isotopic composition; SFCOMPO, initially developed in Japan Atomic Energy research Institute. This paper describes latest version of SFCOMPO and the future development plan in OECD/NEA.

Journal Articles

Development of fission source acceleration method for slow convergence in criticality analyses by using matrix eigenvector applicable to spent fuel transport cask with axial burnup profile

Kuroishi, Takeshi; Nomura, Yasushi

Journal of Nuclear Science and Technology, 40(6), p.433 - 440, 2003/06

 Percentile:100(Nuclear Science & Technology)

Effective source acceleration method is studied in criticality safety analysis for realistic spent fuel transport cask. Various axial burnup profiles based on in-core flux measurements are proposed in the OECD/NEA/BUC benchmark Phase II-C. In some cases, calculations by ordinary Monte Carlo method show very slow convergence of fission source distribution, and unacceptably large skipped cycles are needed. The matrix eigenvector calculation that has been developed and incorporated in the ordinary Monte Carlo calculation to improve the slow convergence is applied to the benchmark. The efficiency of this method depends on the precision of matrix elements. In a certain stage of insufficient convergence of fission source distribution, especially for this benchmark of very slow convergence, more acceleration procedure causes anomalous results because of large statistical fluctuations of matrix elements corresponding to low source levels. Therefore, we propose effective source acceleration method with less calculation time than increasing histories for the estimation of matrix elements.

JAEA Reports

Extended calculations of OECD/NEA phase II-C burnup credit criticality benchmark problem for PWR spent fuel transport cask by using MCNP-4B2 code and JENDL-3.2 library

Kuroishi, Takeshi; Hoang, A.; Nomura, Yasushi; Okuno, Hiroshi

JAERI-Tech 2003-021, 60 Pages, 2003/03

JAERI-Tech-2003-021.pdf:4.56MB

The reactivity effect of the asymmetry of axial burnup profile is studied for PWR spent fuel transport cask proposed in OECD/NEA Phase II-C benchmark. The axial burnup profiles are based on in-core flux measurements. Criticality calculations are performed with the continuous energy Monte Carlo code MCNP-4B2 and the nuclear data library JENDL-3.2. Calculations are carried out not only for cases in the benchmark but also for symmetric burnup cases. Both actinide-only approach and actinide plus fission product approach is considered. The end effect is more sensitive to higher burnup asymmetry. The axial fission distribution becomes strongly asymmetric as its peak shifts toward the fuel top end. The peak of fission distribution gets higher with the increase of either the burnup asymmetry or the assembly-averaged burnup. The conservatism of uniform axial burnup assumption for the actinide-only approach is estimated quantitatively in comparison with the keff result calculated with experiment-based strongest asymmetric axial burnup profile for the actinide plus fission product approach.

JAEA Reports

Acceleration of criticality analysis solution convergence by matrix eigenvector for a system with weak neutron interaction

Nomura, Yasushi; Takada, Tomoyuki; Kadotani, Hiroyuki*; Kuroishi, Takeshi

JAERI-Tech 2003-020, 88 Pages, 2003/03

JAERI-Tech-2003-020.pdf:4.31MB

no abstracts in English

JAEA Reports

Derivation of correction factor to be applied for calculated results of BWR fuel isotopic composition by ORIGEN2.1 code

Nomura, Yasushi; Mochizuki, Hiroki*

JAERI-Tech 2002-068, 131 Pages, 2002/11

JAERI-Tech-2002-068.pdf:5.59MB

no abstracts in English

Journal Articles

New acceleration method of source convergence for loosely coupled multi unit system by using matrix K calculation

Kuroishi, Takeshi; Nomura, Yasushi

Proceedings of International Conference on the New Frontiers of Nuclear Technology; Reactor Physics, Safety and High-Performance Computing (PHYSOR 2002) (CD-ROM), 10 Pages, 2002/10

To accelerate the slow convergence of the fission source distribution, the matrix k calculation has been developed and incorporated in the ordinary Monte Carlo method. The acceleration can be performed by the fission source correction using the eigenvector of the fission matrix, if the coupling coefficients are approximately evaluated in the middle of Monte Carlo calculation. In this paper, we propose two effective applications of the matrix k, that is, the acceleration repetition method and the source generation method. The former simply repeats the matrix k calculation, and the result for the irradiated fuel pin cell shows enough effective to accelerate the fission source on the criticality estimation. However, in some cases of the loosely coupled multi unit system, the repetition of matrix k more than twice could not be carried out to get into convergence because of many units of low source level. The latter is newly devised here to apply to such cases. The checkerboard fuel storage rack is one of the typical cases, and the calculated results show the effectiveness of this method.

Journal Articles

Comparison of burnup calculation results using several evaluated nuclear data files

Suyama, Kenya; Katakura, Junichi; Kiyosumi, Takehide*; Kaneko, Toshiyuki*; Nomura, Yasushi

Journal of Nuclear Science and Technology, 39(1), p.82 - 89, 2002/01

 Times Cited Count:8 Percentile:47.26(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Derivation of correction factor to be applied for calculated results of PWR fuel isotopic composition by ORIGEN2 code

Suyama, Kenya; Murazaki, Minoru*; Mochizuki, Hiroki*; Nomura, Yasushi

JAERI-Tech 2001-074, 119 Pages, 2001/11

JAERI-Tech-2001-074.pdf:4.21MB

no abstracts in English

JAEA Reports

Preparation of data relevant to "equivalent uniform burnup" and "equivalent initial enrichment" for burnup credit evaluation

Nomura, Yasushi; Murazaki, Minoru*; Okuno, Hiroshi

JAERI-Data/Code 2001-029, 120 Pages, 2001/11

JAERI-Data-Code-2001-029.pdf:6.16MB

no abstracts in English

Journal Articles

Activities for revising Nuclear Criticality Safety Handbook

Okuno, Hiroshi; Nomura, Yasushi

Proceedings of the 2001 Topical Meeting on Practical Implementation of Nuclear Criticality Safety (CD-ROM), 8 Pages, 2001/11

The nuclear criticality safety handbook of Japan was first published in 1988, which was translated into English in 1995. This paper intends to introduce the American community of nuclear criticality safety to activities for revising the Japanese handbook, putting an emphasis on practical use of code validation. They include (1) publication of "Nuclear Criticality Safety Handbook, Version 2" and its English translation, (2) publication of "A Guide Introducing Burnup Credit, Preliminary Version," and (3) preparation of "Nuclear Criticality Safety Handbook, -Data Collection-, Version 2."

JAEA Reports

Spent fuel composition data base system on WWW; SFCOMPO on WWW Ver.2

Mochizuki, Hiroki*; Suyama, Kenya; Nomura, Yasushi; Okuno, Hiroshi

JAERI-Data/Code 2001-020, 394 Pages, 2001/08

JAERI-Data-Code-2001-020.pdf:17.99MB

no abstracts in English

64 (Records 1-20 displayed on this page)